Zirconium alloy composition for nuclear fuel cladding tube forming protective oxide film, zirconium alloy nuclear fuel cladding tube manufactured using the composition, and method of manufacturing the zirconium alloy nuclear fuel cladding tube

ABSTRACT

Disclosed herein is a zirconium alloy composition for nuclear fuel cladding tubes, comprising: 1.6˜2.0 wt % of Nb; 0.05˜0.14 wt % of Sn; 0.02˜0.2 wt % of one or more elements selected from the group consisting of Fe, Cr and Cu; 0.09˜0.15 wt % of O; 0.008˜0.012 wt % of Si; and a balance of Zr, a nuclear fuel cladding tube comprising the zirconium alloy composition, and a method of manufacturing the nuclear fuel cladding tube. Since the nuclear fuel cladding tube made of the zirconium alloy composition can maintain excellent corrosion resistance by forming a protective oxide film thereon under the conditions of high-temperature and high-pressure cooling water and water vapor, it can be usefully used as a nuclear fuel cladding tube for light water reactors or heavy water reactors, thus improving the economical efficiency and safety of the use of nuclear fuel.

CROSS-REFERENCES TO RELATED APPLICATION

This patent application claims the benefit of priority under 35 U.S.C. §119 of Korean Patent Application No. 10-2008-0043446 filed on May 9, 2008, the contents of which are incorporated herein by reference.

BACKGROUND OF THE INVENTION

1. Field of the Invention

The present invention relates to a zirconium alloy composition for nuclear fuel cladding tubes, which can maintain excellent corrosion resistance for a long period of time by forming a protective oxide film under the operating environment of a light water reactor or a heavy water reactor, a zirconium alloy nuclear fuel cladding tube manufactured using the zirconium alloy composition, and a method of manufacturing the zirconium alloy nuclear fuel cladding tube.

2. Description of the Related Art

A nuclear fuel cladding tube used for a nuclear fuel assembly of a light water reactor or a heavy water reactor is predominantly manufactured using a zirconium alloy. To date, zircaloy-2 comprising 1.2˜1.7 wt % of tin (Sn), 0.07˜0.2 wt % of iron (Fe), 0.05˜1.15 wt % of chromium (Cr), 0.03˜0.08 wt % of nickel (Ni), 900˜1500 ppm of oxygen (O) and a balance of zirconium (Zr), and zircaloy-4 comprising 1.20˜1.70 wt % of tin (Sn), 0.18˜0.24 wt % of iron (Fe), 0.07˜1.13 wt % of chromium (Cr), 0.07 wt % or less of nickel (Ni), 900˜1500 ppm of oxygen (O) and a balance of zirconium (Zr) have been most widely used in the manufacture of nuclear fuel cladding tubes.

However, recently in order to improve the economical efficiency of a nuclear reactor, a high burn-up operation, in which the burning period of nuclear fuel increases, has been employed. Due to the high burn-up operation, the time for reacting nuclear fuel with high-temperature and high-pressure cooling water and water vapor has increased. Therefore, when the conventional Zircaloy-2 or Zircaloy-4 is used as a raw material for a nuclear fuel cladding tube, there is a problem in that the corrosion phenomenon of the nuclear fuel becomes serious.

Therefore, research into developing materials which can be used to manufacture high burnup nuclear fuel cladding tubes having excellent corrosion resistance to high-temperature and high-pressure cooling water and water vapor is being conducted. Thus, nuclear fuel cladding tubes having better performance than the conventional nuclear fuel cladding tubes made of zircalloy-2 or zircalloy-4 are being developed.

As such, newly-developed zirconium alloy nuclear fuel cladding tubes are most characterized by the fact that they contain niobium (Nb) to improve corrosion resistance. However, the change in the corrosion resistance of the Nb-containing zirconium alloy nuclear fuel cladding tubes is very sensitive to the kind and amount of added elements, the size and distribution of precipitates present in a microstructure, and the like. Accordingly, in order to manufacture a Nb-containing zirconium alloy nuclear fuel cladding tube having excellent corrosion resistance, it is most important to optimize the kind and amount of the elements added to a zirconium alloy.

Conventional technologies related to Nb-containing zirconium alloys used to manufacture nuclear fuel cladding tubes for nuclear reactors are described as follows.

U.S. Pat. No. 5,838,753 and European Patent No. 1,111,623 disclose a method of preparing a zirconium alloy for nuclear fuel cladding tubes and structural parts for high burnup, in which the zirconium alloy comprises 0.5˜3.25 wt % of niobium (Nb) and 0.3˜1.8 wt % of tin (Sn). More specifically, these patents disclose a method of manufacturing a nuclear fuel cladding tube, comprising the steps of: heating a zirconium alloy billet to a temperature above 950° C. and then rapidly quenching the billet to a temperature below the α-transformation temperature in the (α+β)-phase range; extruding the quenched billet at a temperature below 600° C. to form a hollow billet; annealing the hollow billet by heating at a temperature up to 590° C.; pilgering the annealed hollow billet; and finally annealing the pilgered hollow billet at a temperature up to 590° C. to form the nuclear fuel cladding tube. Here, the zirconium alloy has a microstructure of β-niobium second phase precipitates distributed uniformly and intergranularly forming radiation resistant second phase precipitates in the alloy matrix so as to result in an increased resistance to aqueous corrosion compared to that of Zircaloy when irradiated to high fluence.

International Patent Publication No. 2001-061062 discloses a method of manufacturing a nuclear fuel cladding tube including 0.6˜2 wt % of niobium (Nb) and a small amount of tin (Sn), in which the ratio of Sn/Fe is 0.25/0.5, 0.4/(0.35˜0.5) or 0.5/(0.25˜0.5), and the amount of Sn+Fe is 0.75 wt % or more. This method comprises the steps of vacuum melting, forging, hot/cold rolling and heat treatment. The final target of the method is to uniformly distribute small-sized β-niobium (β-Nb) precipitates and zirconium-niobium-iron (Zr—Nb—Fe) precipitates in the zirconium alloy matrix.

Japanese Patent Publication No. 2001-208879 discloses a nuclear fuel assembly member having a welding area. Here, a zirconium alloy or a zircalloy including 0.2˜1.5 wt % of niobium (Nb) is annealed at a temperature of 400˜620° C. in order to improve the corrosion resistance of the welding area of the nuclear fuel assembly member.

International Patent Publication Nos. 2001-024193 and 2001-024194 disclose a zirconium alloy for nuclear reactor components, comprising 0.02˜1 wt % of iron (Fe), 0.8˜2.3 wt % of niobium (Nb), 2000 ppm or less of tin (Sn), 2000 ppm or less of oxygen (O), 100 ppm or less of carbon (C), 5˜35 ppm of sulfur (S), and 0.25 wt % or less of chromium+vanadium (Cr+V).

As described above, in the field of nuclear fuel cladding tubes, research into making a zirconium alloy nuclear fuel cladding tube having excellent corrosion resistance using a niobium-containing zirconium alloy by changing the kind and amount of the elements added to the zirconium alloy has been continuously conducted. However, in nuclear power plants, in order to meet high-burnup long-period operation conditions, it is still required to develop a zirconium alloy nuclear fuel cladding tube having more excellent corrosion resistance than that of conventional nuclear fuel cladding tubes. When a zirconium alloy nuclear fuel cladding tube is used in the reactor environment, the corrosion phenomenon thereof cannot be avoided, and thus an oxide film is formed on the surface of the zirconium alloy nuclear fuel cladding tube. Although it is well known that the corrosion resistance of a nuclear fuel cladding tube is determined by the protection ability of an oxide film formed in the early stage of corrosion, to date, zirconium alloys for nuclear fuel cladding tubes have been developed based on only experimental results, and technologies for improving the protection ability of the oxide film have not yet been developed.

Therefore, while the present inventors were conducting research into niobium-containing zirconium alloy nuclear fuel cladding tubes having excellent corrosion resistance, they developed a zirconium alloy composition which can greatly improve the corrosion resistance of nuclear fuel cladding tubes by forming a protective oxide film thereon in the reactor environment, thereby completing the present invention.

SUMMARY OF THE INVENTION

Accordingly, the present invention has been made to solve the above-mentioned problems, and an object of the present invention is to provide a zirconium alloy composition for nuclear fuel cladding tubes, which can maintain excellent corrosion resistance for a long period time by forming a protective oxide film under the operating environment of a light water reactor or a heavy water reactor.

Another object of the present invention is to provide a zirconium alloy nuclear fuel cladding tube having excellent corrosion resistance, which can meet high-burnup long-period operation conditions, and a method of manufacturing the same.

In order to accomplish the above objects, the present invention provides a zirconium alloy composition for nuclear fuel cladding tubes, comprising: 1.6˜2.0 wt % of Nb; 0.05˜0.14 wt % of Sn; 0.02˜0.2 wt % of one or two elements selected from the group consisting of Fe, Cr and Cu; 0.09˜0.15 wt % of 0; 0.008˜0.012 wt % of Si; and a balance of Zr, a nuclear fuel cladding tube comprising the zirconium alloy composition, and a method of manufacturing the nuclear fuel cladding tube.

BRIEF DESCRIPTION OF THE DRAWINGS

The above and other objects, features and advantages of the present invention will be more clearly understood from the following detailed description taken in conjunction with the accompanying drawing, in which:

FIG. 1 is a graph showing the weight gain of the nuclear fuel cladding tube, which was corroded for 480 days, according to an embodiment of the present invention.

DESCRIPTION OF THE PREFERRED EMBODIMENTS

Hereinafter, preferred embodiments of the present invention will be described in detail with reference to the attached drawings.

The present invention provides a zirconium alloy composition for nuclear fuel cladding tubes, including: 1.6˜2.0 wt % of Nb; 0.05˜0.14 wt % of Sn; 0.02˜0.2 wt % of one or two elements selected from the group consisting of Fe, Cr and Cu; 0.09˜0.15 wt % of O; 0.008˜0.012 wt % of Si; and a balance of Zr.

Here, the zirconium alloy composition may include: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.05 wt % of one element selected from the group consisting of Fe, Cr and Cu; 0.12 wt % of O; 0.01 wt % of Si; and a balance of Zr.

Further, the present invention provides a zirconium alloy composition for nuclear fuel cladding tubes, including: 1.6˜2.0 wt % of Nb; 0.05˜0.14 wt % of Sn; 0.02˜0.2 wt % of Fe; 0.02˜0.2 wt % of Cr or Cu; 0.09˜0.15 wt % of O; 0.008˜0.012 wt % of Si; and a balance of Zr.

Here, the zirconium alloy composition may include: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.05 wt % of Fe; 0.05 wt % of Cr or Cu; 0.12 wt % of O; 0.01 wt % of Si; and a balance of Zr.

The zirconium alloy composition may include: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.1 wt % of Fe; 0.1 wt % of Cr or Cu; 0.12 wt % of O; 0.01 wt % of Si; and a balance of Zr.

The zirconium alloy composition may include: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.2 wt % of Fe; 0.1 wt % of Cr or Cu; 0.12 wt % of O; 0.01 wt % of Si; and a balance of Zr.

Generally, it is known that an oxide film, which is formed when a zirconium alloy nuclear fuel cladding tube is corroded under a reactor environment, has a lamellar structure. The reason why the oxide film has a lamellar structure is that crystal grains are periodically and repetitively changed from columnar grains to equiaxial grains when the oxide film is grown. This periodical and repetitive procedure is referred to as ‘transition’. A protective oxide film is generally characterized in that the thickness of each layer of its lamellar structure is large, and the length and width of columnar grains constituting each layer are large. When this oxide film having such characteristics is formed in the early stage of the corrosion of a nuclear fuel cladding tube, the diffusion rate of oxygen ions through the oxide film is decreased, so that the growth rate of the oxide film is also decreased, thereby improving the corrosion resistance of the nuclear fuel cladding tube.

The conditions that are required to form a protective oxide film on a nuclear fuel cladding tube are determined by the kind and composition ratio of the elements constituting a zirconium alloy. Therefore, hereinafter, the characteristics and composition ratios of the elements constituting a zirconium alloy composition for nuclear fuel cladding tubes according to the present invention will be described in detail in terms of the formation of a protective oxide film.

(1) Niobium (Nb)

Niobium (Nb) is known as an element for stabilizing the beta (β) phase of zirconium. The effects of niobium (Nb) influencing corrosion are different from each other in results. Generally, it is known that when the amount of niobium (Nb) added to a zirconium alloy composition is less than 0.5 wt % (low niobium content), the corrosion resistance of the zirconium alloy is greatly increased and the workability thereof is improved, whereas, when the amount of niobium (Nb) added to the zirconium alloy composition is more than 1.0 wt % (high niobium content), the corrosion resistance of the zirconium alloy are increased. When niobium (Nb) is added to a zirconium matrix in more than solid solubility, solid solutions and precipitates are formed in the zirconium matrix, thereby improving the mechanical properties of zirconium.

It is generally accepted that the addition of niobium (Nb) contributes greatly to the increase in corrosion resistance of a zirconium alloy. Actually, several kinds of niobium-containing zirconium alloy nuclear fuel cladding tubes were developed, and have been practically employed as nuclear fuel cladding tubes in nuclear power plants. However, the causes of the improvement in corrosion resistance of a niobium-containing zirconium alloy have not yet been clearly verified. Several causes of the increase in corrosion resistance of the niobium-containing zirconium alloy have been proposed, and among them, it is known that the strongest contributor to the increase in corrosion resistance thereof is β-niobium precipitates, and that the corrosion resistance thereof improves as the size of the β-niobium precipitates decreases. Sine the corrosion resistance of the zirconium alloy is influenced by the characteristics of the oxide film formed in the early stage of the corrosion of a nuclear fuel cladding tube, it improves if the oxide film maintains the protective properties against the diffusion of oxygen for a long period of time. From this viewpoint, fine and uniform β-niobium precipitates are advantageous to the improvement in the corrosion resistance of the zirconium alloy because they can make the internal stress of the oxide film uniform and can maintain the protective properties of the oxide film for a long period of time even when they penetrate into the oxide film through the corrosion process of the zirconium alloy. For this reason, the amount of niobium (Nb) may be 1.6˜2.0 wt %, preferably 1.7˜1.9 wt %.

(2) Tin (Sn)

Tin (Sn) is known as an element for stabilizing the alpha (α) phase of zirconium, and serves to improve the mechanical strength of a zirconium alloy by solid-solution hardening. It is known that the crystal grain size of an oxide film is decreased by the addition of tin (Sn). Further, decreasing the amount of tin (Sn) added to a zirconium alloy composition advantageously aids the formation of a protective oxide film, but when tin (Sn) is not added to the zirconium alloy composition at all, the corrosion rate of the zirconium alloy can be rapidly accelerated in LiOH corrosion conditions. Accordingly, in the present invention, it is preferred that the amount of tin (Sn) be 0.05˜0.14 wt % as long as it does not greatly influence the decrease in corrosion resistance of the zirconium alloy.

(3) Iron (Fe), Chromium (Cr) and Copper (Cu) (Transition Metal Elements)

Transition metals, such as iron (Fe), chromium (Cr), copper (Cu) and the like, make the growth of an oxide film irregular. However, due to this phenomenon, it is possible to prevent the oxide film from growing in only one direction, so that it is possible to prevent the oxide film from being abruptly destroyed. Further, it is known that the deformability of the oxide film is improved by the addition of these transition metals. However, when the amount of the transition metal added to a zirconium alloy composition is increased, the workability of the zirconium alloy at the time of manufacturing a nuclear fuel cladding tube is decreased. Therefore, in the present invention, the amount of the transition metal may be 0.02˜0.2 wt %, preferably 0.05˜0.1 wt %, and the transition metal may be one or two elements selected from among iron (Fe), chromium (Cr) and copper (Cu).

(4) Silicon (Si) and Oxygen (O)

Silicon (Si) serves to decrease the absorptivity of hydrogen in a zirconium matrix and to delay the transition phenomenon in which the corrosion rate of a zirconium alloy rapidly increases as time advances, and oxygen (O) serves to improve the mechanical strength of the zirconium alloy because it dissolves in the zirconium matrix and thus causes solid-solution hardening.

In the zirconium alloy composition according to the present invention, the amount of silicon (Si), which is added to the zirconium alloy composition in very small quantities, may be 0.008˜0.012 wt %, preferably 0.009˜0.011 wt %. Further, the amount of oxygen (O) added thereto may be 0.09˜0.15 wt %, preferably 0.11˜0.13 wt %. When the amount of silicon (Si) deviates from this range, the corrosion resistance of the zirconium alloy can be decreased, and when the amount of oxygen (O) deviates from this range, the corrosion resistance of the zirconium alloy can be decreased and the workability thereof can be deteriorated.

In addition, the present invention provides a zirconium alloy nuclear fuel cladding tube manufactured by the above-mentioned zirconium alloy composition for nuclear fuel cladding tubes.

Further, the present invention provides a method manufacturing a zirconium alloy nuclear fuel cladding tube.

The method of manufacturing a zirconium alloy nuclear fuel cladding tube according to the present invention includes the steps of: 1) vacuum-arc-remelting and cooling a mixture of the elements constituting the zirconium alloy composition to form an ingot; 2) forging the ingot at a temperature of 1000˜1200° C.; 3) solution-heat-treating the forged ingot at a temperature of 1000˜1200° C. for 10˜40 minutes and then cooling the solution-heat-treated ingot to a temperature of 300˜400° C. at a cooling rate of 300˜400° C./s; 4) extruding the cooled ingot at a temperature of 600˜640° C. to form an extruded shell; 5) primarily heat-treating the extruded shell at a temperature of 570˜610° C. for 2˜4 hours; 6) cold-working the primarily heat-treated extruded shell 2˜5 times and intermediately heat-treating it 1˜4 times between the cold working processes at a temperature of 570˜610° C. for 3˜10 hours to prepare a zirconium alloy nuclear fuel cladding tube; and 7) finally heat-treating the prepared zirconium alloy nuclear fuel cladding tube at a temperature of 470˜580° C. for 1˜100 hours to manufacture a zirconium alloy nuclear fuel cladding tube.

The steps of the method of manufacturing a zirconium alloy nuclear fuel cladding tube according to the present invention will be described in detail.

First, in step 1, an ingot is formed by mixing the elements constituting the zirconium alloy composition according to the present invention in a predetermined range and then vacuum-arc-remelting and cooling the mixture.

The ingot may be formed through a vacuum arc remelting (VAR) process. Specifically, a chamber is maintained in a vacuum of 1×10⁻⁵ torr, argon (Ar) gas is injected into the chamber at a pressure of 0.1˜0.3 torr, and then an electric current of 500˜1000 A is applied to the chamber to melt the mixture, and then the melted mixture is cooled to form the ingot in the form such as button. In this case, in order to prevent the segregation of impurities or the nonuniform distribution of the zirconium alloy composition in the button, the mixture may be repetitively remelted 3˜6 times. In the cooling process, in order to prevent the oxidization of the remelted mixture, it may be cooled while injecting an inert gas.

Next, in step 2, the ingot formed in step 1 is forged in a β-phase region. In this step, in order to destroy the cast structure in the ingot formed in step 1, the ingot may be forged in the β-phase region at a temperature of 1000˜1200° C. When the forging temperature of the ingot is below 1000° C., there is a problem in that the cast structure in the ingot cannot be easily destroyed, and when the forging temperature thereof is above 1200 there is a problem in that the heat treatment cost is increased.

Next, step 3 is a β-quenching step of solution-heat-treating the ingot forged in step 2 in the β-phase region and then rapidly cooling the solution-heat-treated ingot. In this step, in order to uniformize the zirconium alloy composition in the ingot and obtain fine precipitates, the forged ingot is solution-heat-treated in the β-phase region and then rapidly cooled. In this case, in order to prevent the oxidization of the forged ingot, the forged ingot may be sealed with a stainless steel sheet and then solution-heat-treated at a temperature of 1000˜1200° C., preferably 1050˜1100° C., for 10˜40 minutes, preferably 20˜30 minutes. After the solution-heat-treatment of the forged ingot, the solution-heat-treated ingot may be cooled using water in the β-phase region to a temperature of 300˜400° C. at a cooling rate of 300˜400° C./s.

Next, step 4 is a hot-working step of extruding the ingot cooled in step 3. In this step, the ingot cooled in step 3 is processed into a hollow billet, and then the hollow billet is hot-extruded to form an extruded shell suitable for cold working. In this case, the extrusion temperature may be 600˜640° C., preferably 625˜635° C. When the extrusion temperature deviates from this range, it is difficult to obtain an extruded shell suitable for the following step 5.

Next, in step 5, the extruded shell formed in step 4 is primarily heat-treated. Specifically, the extruded shell may be heat-treated at a temperature of 570˜610° C. for 2˜4 hours, preferably at a temperature of 575˜585° C. for 2.5˜3.5 hours.

Next, in step 6, the extruded shell primarily heat-treated in step 5 is repetitively cold-worked and intermediately heat-treated several times to prepare a zirconium alloy nuclear fuel cladding tube. In this step, the cold-working and intermediate heat treatment of the primarily heat-treated extruded shell may be formed by cold-working the primarily heat-treated extruded shell 2˜5 times and intermediately heat-treating it 1˜4 times between the cold working process. In this case, the intermediate heat treatment of the extruded shell may be performed at a temperature of 570˜610° C. for 3˜10 hours. The reason for repetitively performing the cold working and intermediate heat treatment is to make a recrystallization texture in a nuclear fuel cladding tube, to finely and uniformly distribute β-niobium precipitates and to bring the concentration of niobium in the zirconium matrix to an equilibrium concentration of 0.3˜0.6 wt %. Through the above processes, a nuclear fuel cladding tube having an outer diameter of 9.5 mm and a thickness of 0.57 mm can be prepared.

Finally, in step 7, the zirconium alloy nuclear fuel cladding tube prepared in step 6 is finally heat-treated. In this case, the final heat treatment of the prepared zirconium alloy nuclear fuel cladding tube may be performed at a temperature of 470˜580° C. for 1˜100 hours. Due to the final heat treatment, the concentration of niobium in an α-zirconium matrix of the zirconium alloy nuclear fuel cladding tube becomes 0.3˜0.6 wt %, and precipitates including β-niobium precipitates are formed. A zirconium alloy including the β-niobium precipitates is required to be heat-treated for a long period of time in order to form an equilibrium texture therein through the final heat treatment. However, in this case, since the corrosion resistance of the zirconium alloy can be deteriorated due to the increase in the size of the precipitates, the average particle size of the β-niobium precipitates may be controlled to be 70 nm or less.

The zirconium alloy nuclear fuel cladding tube according to the present invention can maintain excellent corrosion resistance under the condition of high-burnup operation, so that it can be usefully used as a nuclear fuel cladding tube for light water reactors or heavy water reactors, thus improving the economical efficiency and safety of nuclear fuel.

Hereinafter, the present invention will be described in more detail with reference to the following Examples. Here, the following Examples are set forth to illustrate the present invention, and the scope of the present invention is not limited thereto.

Example 1 Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube

(1) Formation of an Ingot

An ingot was formed using a zirconium alloy composition including 1.8 wt % of Nb, 0.1 wt % of Sn, 0.05 wt % of Fe, 0.12 wt % of O, 0.01 wt % of Si and a balance of zirconium through a vacuum arc remelting (VAR) process. Here, reactor-grade sponge zirconium defined clearly in the ASTM B349 was used as the balance of zirconium, and the elements constituting the zirconium alloy composition had a high purity of 99.99%. Further, silicon (Si), oxygen (O) and the sponge zirconium were primarily melted to prepare a mother alloy, and then a desired amount of the mother alloy was added at the time of remelting the ingot. In this case, in order to prevent the segregation of impurities or the nonuniform distribution of the zirconium alloy composition, the zirconium alloy composition was repetitively remelted 4 times. Further, in order to prevent the oxidization of the remelted zirconium alloy composition, a chamber was maintained in a vacuum of 1×10⁻⁵ torr, argon (Ar) gas having high purity (99.99%) was injected into the chamber and then an electric current of 500 A was applied to the chamber to melt the mixture, and then the melted mixture was cooled to form the ingot in a water-cooled copper crucible having a diameter of 60 mm and including cooling water having a pressure of 1 kgf/cm².

(2) Forging of an Ingot in a β-Phase Region

In order to destroy the cast structure in the ingot, the ingot was forged in the β-phase region at a temperature of 1100° C.

(3) β-Quenching

In order to uniformize the zirconium alloy composition in the ingot, the ingot forged was solution-heat-treated in the β-phase region at a temperature of 1050° C. for 20 minutes. Further, in order to prevent the oxidization of the ingot, the ingot was cladded with a stainless steel sheet having a thickness of 1 mm. After the solution heat treatment, the ingot was rapidly cooled to a temperature of 400° C. or less at a cooling rate of 300° C./s or more to form a martensite structure or a Widmanstatten structure. Thereafter, in order to remove moisture remaining in the cladded ingot, the cooled ingot was dried at a temperature of 150 for 24 hours.

(4) Hot Working

The ingot β-quenched was processed into a hollow billet, and then the hollow billet was hot-extruded at a temperature of 630° C. to form an extruded shell suitable for cold working.

(5) Primary Heat Treatment

The extruded shell formed was primarily heat-treated at a temperature of 580° C. for 3 hours.

(6) Cold Working and Intermediate Heat Treatment

The extruded shell primarily heat-treated was primarily cold-worked and then intermediately heat-treated in vacuum at a temperature of 580° C. for 2 hours. Subsequently, the extruded shell was secondarily cold-worked and then intermediately heat-treated in a vacuum at a temperature of 580° C. for 2 hours. Then, the extruded shell was tertiarily cold-worked and then intermediately heat-treated in vacuum at a temperature of 580° C. for 2 hours. Thereafter, the extruded shell was finally cold-worked to prepare a nuclear fuel cladding tube having an outer diameter of 9.5 mm and a thickness of 0.57 mm.

(7) Final Heat Treatment

The nuclear fuel cladding tube prepared was finally heat-treated in a vacuum at a temperature of 470˜580° C. for 10 hours to manufacture a zirconium alloy nuclear fuel cladding tube.

Example 2 Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube

A zirconium alloy nuclear fuel cladding tube was manufactured in the same manner as in Example 1, except that the zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of Sn, 0.05 wt % of Cr, 0.12 wt % of O, 0.01 wt % of Si and a balance of zirconium.

Example 3 Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube

A zirconium alloy nuclear fuel cladding tube was manufactured in the same manner as in Example 1, except that the zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of Sn, 0.05 wt % of Cu, 0.12 wt % of O, 0.01 wt % of Si and a balance of zirconium.

Example 4 Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube

A zirconium alloy nuclear fuel cladding tube was manufactured in the same manner as in Example 1, except that the zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of Sn, 0.05 wt % of Fe, 0.05 wt % of Cr, 0.12 wt % of O, 0.01 wt % of Si and a balance of zirconium.

Example 5 Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube

A zirconium alloy nuclear fuel cladding tube was manufactured in the same manner as in Example 1, except that the zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of Sn, 0.2 wt % of Fe, 0.1 wt % of Cr, 0.12 wt % of O, 0.01 wt % of Si and a balance of zirconium.

Example 6 Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube

A zirconium alloy nuclear fuel cladding tube was manufactured in the same manner as in Example 1, except that the zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of Sn, 0.05 wt % of Fe, 0.05 wt % of Cu, 0.12 wt % of O, 0.01 wt % of Si and a balance of zirconium.

Example 7 Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube

A zirconium alloy nuclear fuel cladding tube was manufactured in the same manner as in Example 1, except that the zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of Sn, 0.1 wt % of Fe, 0.1 wt % of Cu, 0.12 wt % of O, 0.01 wt % of Si and a balance of zirconium.

Comparative Example 1

A zirconium alloy nuclear fuel cladding tube was manufactured in the same manner as in Example 1, except that the zirconium alloy composition includes 1.3 wt % of Sn, 0.2 wt % of Fe, 0.1 wt % of Cr, 0.12 wt % of O, 0.01 wt % of Si and a balance of zirconium.

Comparative Example 2

A zirconium alloy nuclear fuel cladding tube was manufactured in the same manner as in Example 1, except that the zirconium alloy composition includes 1.0 wt % of Nb, 1.0 wt % of Sn, 0.1 wt % of Fe, 0.12 wt % of O, 0.01 wt % of Si and a balance of zirconium.

Comparative Example 3

A zirconium alloy nuclear fuel cladding tube was manufactured in the same manner as in Example 1, except that the zirconium alloy composition includes 1.0 wt % of Nb, 0.12 wt % of 0, 0.01 wt % of Si and a balance of zirconium.

The above-mentioned zirconium alloy compositions are shown in Table 1.

TABLE 1 Nb Sn Fe Cr Cu O Si Class. (wt %) (wt %) (wt %) (wt %) (wt %) (wt %) (wt %) Zr Exp. 1 1.8 0.1 0.05 0.12 0.01 balance Exp. 2 1.8 0.1 0.05 0.12 0.01 balance Exp. 3 1.8 0.1 0.05 0.12 0.01 balance Exp. 4 1.8 0.1 0.05 0.05 0.12 0.01 balance Exp. 5 1.8 0.1 0.2 0.1 0.12 0.01 balance Exp. 6 1.8 0.1 0.05 0.05 0.12 0.01 balance Exp. 7 1.8 0.1 0.1 0.1 0.12 0.01 balance Comp. Exp. 1 1.3 0.2 0.1 0.12 0.01 balance Comp. Exp. 2 1.0 1.0 0.1 0.12 0.01 balance Comp. Exp. 3 1.0 0.12 0.01 balance

Experimental Example 1 Corrosion Experiments for Evaluating Corrosion Resistance

Corrosion experiments for evaluating the corrosion resistance of the nuclear fuel cladding tubes manufactured according to the present invention were conducted as follows.

The nuclear fuel cladding tubes manufactured in Examples 1 to 7 and Comparative Examples 1 to 3 were fabricated into corrosion specimens having a length of 50 mm, and then the corrosion specimens were immersed into a mixed solution of water (H₂O), nitric acid (HNO₃) and hydrofluoric acid (HF) having H₂O:HNO₃:HF=50:40:10 (v/v) to remove impurities and fine defects from the surfaces of the corrosion specimens. Before the surface-treated corrosion specimens were put into a corrosion test apparatus, the surface areas and initial weights thereof were measured. Subsequently, the corrosion specimens were put into the corrosion test apparatus simulating a reactor environment using water conditions of 360° C. (18.9 MPa) and then corroded for 90 and 480 days, and then the increases in weights of the corrosion specimens were measured, so that the weight gains per surface area of the corrosion specimens were calculated, thereby quantitatively evaluating the corrosion degree of the corrosion specimens. The corrosion test results are shown in Table 2 and FIG. 1.

TABLE 2 Weight gain (mg/dm²) Class. 90 days 480 days Exp. 1 26.64 47.51 Exp. 2 28.66 47.60 Exp. 3 28.18 47.08 Exp. 4 29.12 49.96 Exp. 5 27.16 61.06 Exp. 6 27.25 46.98 Exp. 7 27.49 48.80 Comp. Exp. 1 28.38 144.17 Comp. Exp. 2 28.84 126.93 Comp. Exp. 3 28.37 62.25

As shown in Table 2 and FIG. 1, after 90 days, the weight gains of the nuclear fuel cladding tubes manufactured in Examples 1 to 7 are in a range of 26.64˜29.12 mg/dm², which differ little from the weight gains (28.37˜28.84 mg/dm²) of the nuclear fuel cladding tubes manufactured in Comparative Examples 1 to 3. However, after 480 days, the weight gains of the nuclear fuel cladding tubes manufactured in Examples 1 to 7 are in a range of 47.08˜61.06 mg/dm², which are 50% or more lower than the weight gains (62.25˜144.17 mg/dm²) of the nuclear fuel cladding tubes manufactured in Comparative Examples 1 to 3.

Experimental Example 2 Observation of an Oxide Film Formed After the Corrosion of a Nuclear Fuel Cladding Tubes

In order to evaluate the ability to protect against corrosion of an oxide film formed on the nuclear fuel cladding tube manufactured according to the present invention, the oxide film, which was formed after the corrosion test, was observed using a transmission electron microscope.

Specimens for observing the oxide film were made thin to such a degree that they could be observed by the transmission electron microscope using ion beams after the corrosion specimens of Examples 1 to 7 and Comparative Examples 1 to 3, which had been corroded for 90 days under a reactor atmosphere, were cut to a thickness of 100 μm. The oxide film specimens were observed by the transmission electron microscope, and thus the crystal grain size distribution of the oxide film was measured, thereby evaluating the ability of the oxide film to protect against corrosion. The results thereof are shown in Table 3.

TABLE 3 Fraction of Length of Width of columnar grain columnar grain columnar grain Class. (%) (nm) (nm) Exp. 1 ~70 ~500 ~60 Exp. 2 ~70 ~500 ~60 Exp. 3 ~70 ~500 ~60 Exp. 4 ~70 ~500 ~60 Exp. 5 ~70 ~500 ~60 Exp. 6 ~70 ~500 ~60 Exp. 7 ~70 ~500 ~60 Comp. Exp. 1 ~20 ~150 ~20 Comp. Exp. 2 ~40 ~250 ~30 Comp. Exp. 3 ~50 ~350 ~40

The corrosion resistance of the nuclear fuel cladding tube is determined by the protective ability of the oxide film formed in the early stage of the corrosion of the nuclear fuel cladding tube. As shown in Table 3, since the oxide films formed on the nuclear fuel cladding tubes manufactured in Examples 1 to 7 are protective oxide films having a larger columnar grain fraction, length and width than those of the oxide films formed on the nuclear fuel cladding tubes manufactured in Comparative Examples 1 to 3, the diffusion of oxygen through an oxide film can be prevented, thereby maintaining excellent corrosion resistance.

As described above, since the nuclear fuel cladding tube made of the zirconium alloy composition according to the present invention can maintain excellent corrosion resistance by forming a protective oxide film thereon under the conditions of high-temperature and high-pressure cooling water and water vapor, it can be usefully used as a nuclear fuel cladding tube for light water reactors or heavy water reactors, thus improving the economical efficiency and safety of the use of nuclear fuel.

Although the preferred embodiments of the present invention have been disclosed for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and substitutions are possible, without departing from the scope and spirit of the invention as disclosed in the accompanying claims. 

1. A zirconium alloy composition for nuclear fuel cladding tubes, comprising: 1.6˜2.0 wt % of Nb; 0.05˜0.14 wt % of Sn; 0.02˜0.2 wt % of one or two elements selected from the group consisting of Fe, Cr and Cu; 0.09˜0.15 wt % of O; 0.008˜0.012 wt % of Si; and a balance of Zr.
 2. The zirconium alloy composition according to claim 1, wherein the composition comprises: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.05 wt % of one element selected from the group consisting of Fe, Cr and Cu; 0.12 wt % of O; 0.01 wt % of Si; and a balance of Zr.
 3. The zirconium alloy composition according to claim 1, wherein the composition comprises: 1.6˜2.0 wt % of Nb; 0.05˜0.14 wt % of Sn; 0.02˜0.2 wt % of Fe; 0.02˜0.2 wt % of Cr or Cu; 0.09˜0.15 wt % of O; 0.008˜0.012 wt % of Si; and a balance of Zr.
 4. The zirconium alloy composition according to claim 3, wherein the composition comprises: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.05 wt % of Fe; 0.05 wt % of Cr or Cu; 0.12 wt % of O; 0.01 wt % of Si; and a balance of Zr.
 5. The zirconium alloy composition according to claim 3, wherein the composition comprises: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.1 wt % of Fe; 0.1 wt % of Cr or Cu; 0.12 wt % of O; 0.01 wt % of Si; and a balance of Zr.
 6. The zirconium alloy composition according to claim 3, wherein the composition comprises: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.2 wt % of Fe; 0.1 wt % of Cr or Cu; 0.12 wt % of O; 0.01 wt % of Si; and a balance of Zr.
 7. A nuclear fuel cladding tube comprising the zirconium alloy composition of claim
 1. 8. A method of manufacturing a zirconium alloy nuclear fuel cladding tube, comprising: 1) vacuum-arc-remelting and cooling a mixture of the elements constituting the zirconium alloy composition of claim 1 to form an ingot; 2) forging the ingot at a temperature of 1000˜1200° C.; 3) solution-heat-treating the forged ingot at a temperature of 1000˜1200° C. for 10˜40 minutes and then cooling the solution-heat-treated ingot to a temperature of 300˜400° C. at a cooling rate of 300˜400° C./s; 4) extruding the cooled ingot at a temperature of 600˜640° C. to form an extruded shell; 5) primarily heat-treating the extruded shell at a temperature of 570˜610 for 2˜4 hours; 6) cold-working the primarily heat-treated extruded shell 2 times and intermediately heat-treating it 1˜4 times between the cold working processes at a temperature of 570˜610° C. for 3˜10 hours to prepare a zirconium alloy nuclear fuel cladding tube; and 7) finally heat-treating the prepared zirconium alloy nuclear fuel cladding tube at a temperature of 470˜580° C. for 1˜100 hours to manufacture a zirconium alloy nuclear fuel cladding tube.
 9. The method of manufacturing a zirconium alloy nuclear fuel cladding tube according to claim 8, wherein, in step 1, the mixture is repetitively vacuum-arc-remelted 3˜6 times to form an ingot.
 10. The method of manufacturing a zirconium alloy nuclear fuel cladding tube according to claim 9, wherein the mixture is vacuum-arc-remelted and then cooled while injecting an inert gas to form an ingot.
 11. The method of manufacturing a zirconium alloy nuclear fuel cladding tube according to claim 8, wherein, in step 7, the final heat treatment is performed in a vacuum.
 12. The method of manufacturing a zirconium alloy nuclear fuel cladding tube according to claim 8, wherein, in step 7, the final heat treatment is performed such that an average particle size of β-niobium precipitates is controlled to be 70 nm or less. 